2005 International Nuclear Atlantic Conference - INAC 2005 Santos, SP, Brazil, August 28 to September 2, 2005 ASSOCIAÇÃO BRASILEIRA DE ENERGIA NUCLEAR - ABEN ISBN: 85-99141-01-5 Evaluation of FPTRAN module of RELAP/SCDAPSIM Code Using PHEBUS FPT-01 Experiment Eduardo H. R. Honaiser 1 , Wagner José Gomes Pereira 2 and Carlos Henrique Calazans Ribeiro 3 1 Centro Tecnológico da Marinha em São Paulo Av. Professor Lineu Prestes 2242 05508-000 São Paulo, SP [email protected]2 Centro Tecnológico da Marinha em São Paulo Av. Professor Lineu Prestes 2242 05508-000 São Paulo, SP [email protected]3 Centro Tecnológico da Marinha em São Paulo Av. Professor Lineu Prestes 2242 05508-000 São Paulo, SP [email protected]ABSTRACT Fission products transport in the piping system of primary circuits is an important area of severe accident’s study, which has grown in importance for the nuclear community after the events of Three Miles Island-2 (1979) and Chernobyl (1986) accidents. Once released in the flow channels, fission products can condense on the piping walls, nucleate aerosols, which can agglomerate and/or deposit on the piping walls, and be convected by the flow. This grown in importance resulted in several model developments and experiments realizations. These models and experiments vary from phenomenological specific to integral. FPTRAN [1] is a model developed for the program SCDAP/RELAPSIM that calculates all phenomenology related to the fission products transport through the piping system The PHEBUS experiments [2] compose the more complete experimental program ever conducted for the understanding of fission products behavior in Reactor Cooling System and containment, being ideal to assess FPTRAN performance. This paper describes the modeling of the experiment and compares simulation and experimental results. The encountered results can be considered satisfactory, except for iodine. This disagreement for iodine is caused, probably, by a wrong chemical form assumed for iodine. 1. INTRODUCTION FPTRAN [1] is a fission product and structural materials (cladding and control rod materials) transport model developed to calculate the amount of fission products and structural materials deposited over the primary circuit piping system, in the condensed and aerosol form. Aerosol transport model is the Lemon and Clark [2] model (discrete ordinates approach). The model calculates vapor deposition for condensation and adsorption. Aerosol agglomeration and deposition is calculated in the code. Agglomeration by Brownian diffusion, gravitational
Transcript
2005 International Nuclear Atlantic Conference - INAC 2005 Santos, SP, Brazil, August 28 to September 2, 2005 ASSOCIAÇÃO BRASILEIRA DE ENERGIA NUCLEAR - ABEN ISBN: 85-99141-01-5
Evaluation of FPTRAN module of RELAP/SCDAPSIM Code Using PHEBUS FPT-01 Experiment
Eduardo H. R. Honaiser1, Wagner José Gomes Pereira2 and Carlos Henrique Calazans
Ribeiro3
1 Centro Tecnológico da Marinha em São Paulo
Av. Professor Lineu Prestes 2242 05508-000 São Paulo, SP [email protected]
2 Centro Tecnológico da Marinha em São Paulo
Av. Professor Lineu Prestes 2242 05508-000 São Paulo, SP
ABSTRACT Fission products transport in the piping system of primary circuits is an important area of severe accident’s study, which has grown in importance for the nuclear community after the events of Three Miles Island-2 (1979) and Chernobyl (1986) accidents. Once released in the flow channels, fission products can condense on the piping walls, nucleate aerosols, which can agglomerate and/or deposit on the piping walls, and be convected by the flow. This grown in importance resulted in several model developments and experiments realizations. These models and experiments vary from phenomenological specific to integral. FPTRAN [1] is a model developed for the program SCDAP/RELAPSIM that calculates all phenomenology related to the fission products transport through the piping system The PHEBUS experiments [2] compose the more complete experimental program ever conducted for the understanding of fission products behavior in Reactor Cooling System and containment, being ideal to assess FPTRAN performance. This paper describes the modeling of the experiment and compares simulation and experimental results. The encountered results can be considered satisfactory, except for iodine. This disagreement for iodine is caused, probably, by a wrong chemical form assumed for iodine.
1. INTRODUCTION FPTRAN [1] is a fission product and structural materials (cladding and control rod materials) transport model developed to calculate the amount of fission products and structural materials deposited over the primary circuit piping system, in the condensed and aerosol form. Aerosol transport model is the Lemon and Clark [2] model (discrete ordinates approach). The model calculates vapor deposition for condensation and adsorption. Aerosol agglomeration and deposition is calculated in the code. Agglomeration by Brownian diffusion, gravitational
difference, and turbulence are calculated. Regarding aerosol deposition, gravitational settling, thermophoresis, laminar and turbulent diffusion are calculated. The PHEBUS FP program is a Light Water Reactor (LWR) source term research program, undertaken by the "Institut de Protection et de Sûreté Nucléaire" (IPSN) of the "Commissariat à l'Energie Atomique" (CEA). The aim of the experimental program is to study the release, transport and behavior of fission products, structure and control rod materials, from the core of a LWR under severe accident conditions, in a scaled reactor circuit and containment building, in order to verify the validity of, or to improve, the analytical models developed on the basis of separate effects test results. The reduced volumetric scale factor compared to a Pressurized Water Reactor (PWR) 900 MW is approximately 1/5000. In this research, the Phebus FPT-01 [3] experiment is presented because there is extensive availability of data, once it is being used in the International Standard Problem (ISP-46). In the subsequent sections are presented the Phebus test facility, the test conduction, and the important lessons learned from the Phebus series of experiments.
2. EXPERIMENTAL DESCRIPTION
2.1. Experimental Facility
The Phebus experimental facility consists of: (1) a driver core and its cooling circuit, (2) a pressurized water loop (PWL), (3) a bundle of fuel rods, (4) an injection loop, (5) an experimental fission product (FP) circuit including a horizontal line and a steam generator, and (6) a containment vessel. Figure 1 presents a simplified diagram of the Phebus fission products circuits. The focus of this article will be on the bundle of rods and the experimental fission product circuit.
Figure 1. Phebus experimental facility
The conditions of the postulated severe accident are applied to a bundle of
representative PWR fuel rods contained inside the test assembly (Figure 2). This is performed in two phases: (a) Pre-irradiation phase, when the bundle is heated out of the phebus reactor
INAC 2005, Santos, SP, Brazil.
to obtain the designed burn-up (~23.4 GWd/tU); (b) Irradiation phase, when the bundle is irradiated in the phebus reactor, with the aid of the pressurized water loop. The heat-up of the fuel bundle is obtained by fission reactions produced by the neutronic flux generated in the driver core. Meanwhile, the steam flow rate at the bundle inlet is adjusted to the prescribed values, through the injection circuit.
The FPT1 test bundle contains 18 fuel rods (UO2), two fresh-instrumented fuel rods introduced into the bundle as well as an Ag, In, Cd (AIC) absorber control rod (containing 80 wt.% silver, 15 wt.% indium and 5 wt.% cadmium). The irradiated fuel rods consist of a stack of fuel pellets housed in a zircaloy (Zry) cladding. The enriched UO2 column is 1 m in length. The instrumented rods are slightly shorter, without the blanket of UO2 pellets in the lower part. The fuel bundle is surrounded by a thoria layer, a ceramic low porosity ZrO2 layer, and an outer dense zirconia coating on the Inconel pressure tube (shroud). Heated metallic sleeves are added above the fuel bundle region in order to maintain outlet gas temperatures at about 700°C (973 K). The spacing between the fuel rods and the control rod in the test assembly is maintained by two zircaloy spacer-grids 43 mm in height located at the 240 and 760 mm elevations. The zircaloy grids also include small inconel springs in contact with the external surface of the fuel rods. In addition, the test bundle contains four zircaloy stiffeners (U tube), located close to the inner surface of the shroud, designed to strengthen the structure of the rod assembly. Figure 2 shows the materials and measures in the fuel bundle region.
Figure 2. Radial Cross Section of the test bundle
The experimental lines are composed successively of: (a) A vertical line directly above the test bundle, leading to a horizontal line (seen in Figure 1); (b) A horizontal pipe, simulating conditions in the hot leg of a PWR primary circuit; (c) a U tube simulating a PWR steam generator; (d) A horizontal line, simulating conditions in the cold leg of a PWR primary circuit, leading to the containment. The vertical line is composed of the upper plenum and the
INAC 2005, Santos, SP, Brazil.
riser. The bottom (0.2 m length) of the upper plenum is unheated and the remainder of the vertical line is regulated to a temperature of 700°C. The horizontal line simulating conditions in the hot leg is made of inconel-600 and its temperature is regulated to 700°C. The steam generator U-tube is made of inconel-600, its downward walls were maintained at 150°C during the test, i.e., above the conditions for steam saturation for the considered pressure. The outlet of the U tube and the horizontal line, simulating conditions in the cold leg, are made of stainless-steel (AISI 304) and were maintained at 150°C. The diameters and lengths of the experimental circuit lines, as well as its materials, are presented in Table 1.
Table 1. Geometrical characteristics of the experimental circuit
Cold Leg 0.03 3.438 Stainless Steel The instrumentation used in the experimental program is divided into nuclear, thermal-hydraulic, and fission products related. The nuclear instrumentation is used to obtain the neutron flux in the test train, and, consequently, the power. The thermal-hydraulic instrumentation is used in the core and circuits measuring thermal-hydraulic parameters. These parameters measurement devices are thermocouples, flow meters and manometers. The fission products instrumentation is used to measure fission products-related parameters, such as aerosol size, composition and mass retained.
2.2. Experiment Conduction The FPT1 test is devoted to studying the phenomenology of severe accident sequences, under oxidizing conditions without steam starvation at low pressure, for which the fission products flow-path involves a portion of the primary circuit, the steam generator and the reactor containment building. The objectives of the test may be stated separately for the bundle and the experimental circuit. The objective for the bundle was to obtain a significant degradation of the fuel rods, in order to maximize the fission products releases. It was estimated, in the pre-test analyses, that about 2 kg of liquefied fuel was necessary to release 70% to 80% of the fission products. The bundle degradation was obtained by reaching the liquefaction temperature of the fuel. It was expected that interactions between the fuel pellets and the molten cladding or control rod could also contribute to the degradation of the bundle. Regarding the circuit, the main objectives were to investigate depletion of the fission products within the primary circuit of the reactor at low pressure (around 0.2 MPa) without steam condensation and to provide data on fission products chemistry including interactions of these fission products with pipe walls at high temperature.
INAC 2005, Santos, SP, Brazil.
The Phebus FPT-1 test can be divided into the preparatory phase, the transition phase, and the bundle-degradation phase. The interest of this research is only the bundle-degradation phase, once it is correspondent to the modeled transient time. The bundle steam pressure was controlled at 0.2 MPa during the bundle degradation phase, and the steam injected at the bottom of the bundle in the subsequent phases of the test was heated to 165°C (438 K) by the pressurized water circuit. The degradation phase of the test was initiated by progressively increasing the nuclear power generated in the fuel rods with a steam flow rate (between 0.5 and 2.2 g/s) injected at the bottom of the bundle under 0.2 MPa. During this phase, lasting 5 hours, the temperatures of the test bundle were continuously controlled so that they followed as closely as possible those foreseen by the pre-test calculations. The bundle degradation phase spanned the period between the experiment beginning time and the reactor shut-down time. The degradation phase of the FPT1 test included: (1) a thermal calibration period, (2) an oxidation period, (3) a power plateau P1, (4) a heat-up period and, (5) a cool-down period. Figure 3 shows the mass flow rate and bundle power history during the several experiment periods.
Figure 3. Bo
A preliminathree-temperC. In the preconstant rateseconds). Thinside the bufor the oxidtemperature The nuclear characterizinpower was
INAC 2005, Santo
Thermal calibration
Pre-oxidation
Oxidation And P1
undary conditions of the FPT-01 experiment
ry period was devoted to the thermal calibration ature plateau, and lasted 7900 seconds. The tempera-oxidation phase, the nuclear power generated in th until 9000 seconds, and the nuclear power was stais power plateau corresponds to a maximum measndle. The pre-oxidation period was set to obtain the ation of the cladding to start to take place signescalation in the bundle.
power was increased again at 9300 seconds up to Pg the oxidation period. In the following phase, the pkept in this level for 23 min (until 14,580 seco
s, SP, Brazil.
Heat-up
of the test bundle including ture has kept constant at 660 e bundle was increased at a
bilized for 5 min (until 9300 ured temperature of 1330°C thermal-hydraulic conditions ificantly, i.e., the onset of
1 = 24 kW (13,200 seconds), ower plateau phase (P1), the nds). During these phases,
significant amount of fission products was released. Also, the rods in the bundle lost their original geometry, occurring extensive oxidation, melt relocation and debris formation. During the subsequent period of the test (the heat-up phase), the temperatures were increased up to fuel liquefaction in order to produce extensive degradation of the bundle and additional fission products releases. Beyond the P1 plateau, the nuclear power was increased at a mean rate of 365 W/min. After 16,500 seconds, the power ramp was reduced to 240 W/min. The test was terminated when several conditions were met, which were indications that the mass of liquefied or molten fuel of 2 Kg was reached. These conditions were obtained from pre-test calculations. The steam mass flow rate was controlled, during the experiment, to maintain a steam-rich environment. Figure 4 contains the more important events in terms of severe accident phenomena, as well as the temperatures in the gas outlet and cladding, and the hydrogen production rate.
Figure 4. Phebus FPT-01 parameters and major events
3. EXPERIMENTAL MODELING The bundle and experimental circuit of Phebus FPT-01 experiment were modeled in SCDAP/RELAP5 code. The nodalization of the test bundle region followed the common sense of the user’s practice, established in the code user’s guideline [4]. Initially, the bundle and the experimental circuit were modeled, and a good temperature agreement in the bundle temperatures was reached. The inlet conditions in the vertical liner were obtained, and then only the experiment circuit was used in the calculation, using the pre-calculated flow history. The flow boundary conditions were set through time dependent volumes and junctions elements. The thermal boundary conditions were set through temperatures histories from thermocouples on the piping walls. The nodes in the piping system have the average length of 400 mm. Figure 5 shows the nodalization diagram of the used model.
INAC 2005, Santos, SP, Brazil.
133 134141 142 143
Cold Leg144
135132
131 Hot Leg 140
130
Steam Generator
Vertical Liner
Figure 5. Nodalization Diagram of the Phebus FPT-01 Experimental Circuit
Two issues were important in the fission products input: speciation and the release model. Chemistry of the fission products, or their speciation, is the most challenging issue of the fission products-behavior study. Speciation has great dependence on the environmental conditions and the time evolution of the environment. The SCDAP/RELAP5 code employs a fixed speciation approach. Additional complications arise from the experimental scenario. The most important is the presence of Re and W in the experimental circuit. Rhenium and tungsten are thermocouple’s materials, which, normally do not exist in actual severe accidents, and are not included in the thermo-chemical database of the code. The effect of tungsten is neglected in the calculation, since its impact in the kinetics of the fission products behavior is not significant, given its volatility and small initial inventory. Rhenium, on the other hand, has a significant impact in the fission products behavior, constituting one of the major components of the aerosol. The behavior of rhenium is similar to silver’s, in terms of vapor pressure. Therefore, silver properties were adopted, when the rhenium property was not available. The selection of the elements to be tracked was based on two criteria. First, the element must have a significant release in the containment. This criterion eliminates from the selection the low volatility elements. The second criterion is the importance of the element. Every element having effect in the composition of the aerosol is considered important. Since no thermo-chemical data for technetium exist in the code, and given its similarity with molybdenum, this element is treated as molybdenum. Antimony (Sb) is not treated, given the difficulty of experimentally measuring it. Rubidium is treated as cesium, given that their chemical behavior is known to be similar. The selection of the species to be used was based on the assumption that most of the elements are transported in the oxide form. The assumed speciation is based on the current knowledge of the most probable species for each element, considering the known oxidizing conditions. The speciation is presented in Table 2.
INAC 2005, Santos, SP, Brazil.
Table 2. Speciation used in the Phebus FPT-01 Relap/Scdap experiment model
Elements Species 1. Cs/Rb CsOH/CsI/ Cs2MoO4 2. I CsI 3. Mo/Tc Mo/ Cs2MoO4 4. Te Te 5. Ag/Re Ag 6. U UO2 7. Cd CdO 8. In In 9. Sn Sn
Release of fission products constitutes the source of the transport. Since the released masses are accurately known, and the time-release behavior is reasonably known, the following procedure was set to model the release of the species in Table 2: (a) Classify the several tracked elements according to its release behavior; (b) Divide the transient period in three phases, with different fission products release behaviors, and determine the percent release of each class of elements, in each phase; (c) Determine a constant release rate for each species, for each phase, considering its classification, and the released amount. The release rates, resultant from these operations, are presented in Table 3.
Table 3. Release rates for element considered in the modeling of FPT-1 experiment
4. COMPARISON OF CALCULATED AND EXPERIMENTAL RESULTS
The comparison of the calculated and experimental results is composed of the following items: (a) Mass balance, where the measured deposited masses of the tracked species in the several circuit parts are compared with the calculated ones; (b) Aerosol composition; (c) Aerosol size. The presentation of the mass balance comparison for the considered elements is divided into four major groups, because of the different behavior and importance of these groups. The
INAC 2005, Santos, SP, Brazil.
nucleating group (Ag and Re) includes the species that nucleate aerosols. The structural materials group includes materials from control rod and cladding (Sn, In and Cd). The fission product species (Cs, Mo, I and Te) are divided into the group that speciation was pre-defined (Cs and Mo) and the group that speciation was not worked (I and Te). Good information about Cs and Mo was available in the Phebus FPT-01 report [3]. For this reason, Cesium transport was calculated in the form of cesium iodide, cesium hydroxide, and cesium molybdate, in the fractions of 0.04, 0,65, and 0,31, respectively. Molybdenum was calculated in the form of cesium molybdate and molybdenum. On the other hand, not much information about speciation for iodine and tellurium is available, resulting in a simple speciation for these elements. The results for Ag and Cd are presented in Figure 6.
48.2
6
1.88
8.31
3.67
11.9
7
4.51
66.6
3
48.4
0
1.30
12.5
0
3.60
65.8
0.00
10.00
20.00
30.00
40.00
50.00
60.00
70.00
80.00
VerticalLine
Hot Leg Sgup SGDown SG Total Cold Leg Total
Ag CalculatedAg Experimental
(a)
27.3
8
0.00
5.40
0.98
6.38 14
.90
48.6
6
23.0
0
0
18.0
0
0.90
41.9
0.00
10.00
20.00
30.00
40.00
50.00
60.00
VerticalLine
Hot Leg Sgup SGDown SG Total Cold Leg Total
Cd CalculatedCd Experimental
(b) Figure 6. Comparison of calculated and experimental deposition fractions (%) for: (a)
Silver; (b) Cadmium
INAC 2005, Santos, SP, Brazil.
For Silver, most of the deposition occurred by condensation on the structural surface on the vertical line. In the other sections of the experimental circuit, deposition occurred basically in the aerosol form. Great amount of deposition occurred on the steam generator up section (8.31 %), due to thermophoresis, as expected (reported in the Phebus ISP-46 report). Considerable deposition was also registered both in the experimental data and in the calculation on the cold leg. All the calculated values are matching the experimental data, within the range of experimental uncertainty. Although, the calculated value was presented dividing the steam generator in two sections, the experimental data in the Phebus report [3] is only presented for the steam generator as a whole. Cadmium was transported part in the oxide form, and part in the elemental form (half-to-half). The cadmium oxide condensed on the vertical leg (about 60%) resulting in a deposition fraction of 28 %, a little higher than the experimental data. In the steam generator, there was deposition of 6% of the cadmium-released mass on the steam generator. Five percent deposited on the steam generator up section (75% in the condensed form), and close to one percent deposited in aerosol form on the steam generator down section. This steam generator total is well below the experimental value. However, a significant amount of elemental cadmium has condensed on the cold leg surface (14 %). Adding up the steam generator depositions with the cold leg deposition, the prediction becomes within the experimental uncertainty width. What probably occurred is that the cadmium actually condensed in the steam generator but, in the calculation, because of errors in the surface temperature or saturation concentration, the condensation was only predicted in the cold leg. The total calculated is a little higher than the experimental data (about 6 %). Figure 7 shows the fission products results for cesium and iodine. Cesium, in the form of cesium molybdate condensed on the vertical line, resulting in a condensation fraction of about 38 % for the element. In the steam generator up section, there was a deposition of cesium molybdate through condensation (about 60 % of the deposited mass) and aerosol deposition, adding up the deposition of 6.9% of the released mass. On the steam generator down section, cesium iodide condensed on the structural surfaces and some cesium molybdate deposited on the aerosol form, resulting in an elemental deposition of 7.7%. In this way, the total predicted mass for cesium in the steam generator has good agreement with the experimental data. Iodine predicted behavior was very different from the experimental observed. Whereas, in the calculation, iodine violently condensed on the steam generator down section, where the saturation conditions for cesium iodide were reached. This is a strong indication that the assumed speciation of iodine is incorrect, once the code predicted the behavior for other elements reasonably well.
INAC 2005, Santos, SP, Brazil.
38.0
3
0.00
6.90
7.72 14
.62
2.07
54.7
2
30.6
4.16 13.5
0
0.71
48.9
7
0.00
10.00
20.00
30.00
40.00
50.00
60.00
VerticalLine
Hot Leg Sgup SGDown SG Total Cold Leg Total
Cs CalculatedCs Experimental
(a)
0.00
0.00
0.00
73.4
8
73.4
8
3.23
76.7
0
0.23
22
0.89
29.2
2
6.1
0.00
10.00
20.00
30.00
40.00
50.00
60.00
70.00
80.00
90.00
VerticalLine
Hot Leg Sgup SGDown SG Total Cold Leg Total
I CalculatedI Experimental
(b) Figure 7. Comparison of calculated and experimental deposition fractions (%) for: (a)
Cesium; (b) Iodine
Figure 8 shows that the aerosol calculated and experimental composition are distinguished by the greater amount of nucleating species in the calculated aerosol structure than in the assumed approximated structure. This greater presence naturally reduced the amount of other species for the calculated composition. Given the high degree of approximation for the assumed experimental structure, and the uncertainties of the experimental result, the code
INAC 2005, Santos, SP, Brazil.
predictions did not present any significant anomaly that deserves analysis. Cadmium and cesium were the species that present the greater difference, but the origin of this difference could be attributed to the degree of approximation, especially given the difficulties in the measurement of these elements described in the Phebus report [3].
The comparison of the aerosol size distributions parameters is presented in Table 4, for the hot leg. The aerosol distribution is assumed as log-normal, and it is characterized the aerosol mean measured diameter (AMMD) and its standard deviation. The aerosol deposition was measured in the hot and cold legs, and, as mentioned in the Phebus report [3], has strong experimental uncertainties. Table 4 shows that the calculated value for the distribution mean value is smaller than the measured. Same behavior was observed in the cold leg. A myriad of causes can be responsible for this discrepancy. For instance, one of the possible causes of this under-prediction is the fact that some important aerosol components (Sn, In) were transported in its elemental form. Since the oxide form has a considerable lower density than its
INAC 2005, Santos, SP, Brazil.
elemental form, the volume would be greater, if they were transported in the oxide form. Hence, speciation also impacts in the aerosol size. Experimental uncertainties can also contribute to the discrepancy. The level of differences, nevertheless, is not capable of affecting the deposition significantly, because the difference on the deposition velocities for these two AMMD (for calculated and experimental) is small.
Table 4. Comparison of log-normal distribution parameters from predicted and experimental data for Ag weight as function of particle diameter
Calc. with synergy
Experimental
AMMD 1.098 1.376 σ 0.365 0.975
5. CONCLUSIONS
FPTRAN model was developed as part of an effort to improve SCDAP/RELAP5 capabilities. This module calculates the amount of fission products and structural materials dynamics in the primary circuit system during a severe accident. Its performance was assessed through the modeling and comparison of the results with the Phebus FPT-01 experimental results. The model has included a typical nodalization for the fission products experimental circuit, and flow boundary conditions were applied through time-dependent elements and thermal boundary conditions were applied through temperatures history, using instrumentation data. In the fission products modeling, the tracked species were selected based on their importance and accuracy of the data. The fission products source term was applied through release rates time-history. During the mass balance comparison with the experimental data, the tracked species were divided in nucleating, structural material, and fission product species. For the nucleating species, the calculated results presented a very good agreement with the experimental. For the structural materials group, some differences in the overall deposited fraction were found. In general, the deposition behavior was well predicted by the code. Same conclusion can be drawn from the cesium calculation in Figure 7 (a). For iodine, on the other hand, the code completely failed to predict the element dynamics. In a preliminary analysis, wrong speciation was considered as the primary cause of this discrepancy, but a deeper analysis need to be done. As a general rule, it is difficult to determine the cause of the disagreement, given the great amount of possible causes and simplifications during the model development and experimental modeling. Aerosol composition comparison has shown that the code performance is reasonable. Uncertainties related to the experimental measurement of the mass fractions do not permit a deeper analysis. Regarding aerosol size, the code has predicted smaller aerosol mean diameter than the experimental. This difference was not capable to strongly affect the aerosol deposition, but an investigation about its cause needs to be addressed.
INAC 2005, Santos, SP, Brazil.
ACKNOWLEDGMENTS To Dr. Chris Alison, from Innovative Systems Software for the support and SCDAP/RELAPSIM usage allowance.
REFERENCES 1. Anghaie, S, and Honaiser, E. “FPTRAN: A Volatile Fission Products and Structural
Materials Transport Code for SCDAP/RELAP5,” Proceedings of the International Conference in Advances for Nuclear Power Plants, ICAPP 2004, Pittsburg, EUA, June, 2004.
2. Institute de Protection et de Surete Nucleaire, “ Phebus FPT-01 Final Report,” CEA Report, Cadarache, France, 2002.
3. Lemmon, E. C., Marwill, E. S., “Aerosol Simulation Including Chemical and Nuclear Reactions,” Proceedings of the Society for Computer Simulation, Chicago, Illinois, 1985.
4. U. S. Nuclear Regulatory Commission, “SCDAP/RELAP5/Mod 3.2- Code Manual Volume III: User’s Guide and Input Manual,” USNRC Report NUREG-CR/6150, Idaho Falls, Idaho, 1995.
